Corrosion is a major concern for reactor structures because in their con­struction many different materials are used which corrode at different rates by electrochemical effect and the corrosion (pitting, cracking, etc.) is accel­erated by neutron radiation. More importantly, the corrosion products from steam generators, piping and other components are transported through the core and deposit on the fuel rods leading to formation of crud, in turn lead­ing to increased fuel temperature and fuel failure. Corrosion can also lead to deposition of radioactive corrosion products on out-of-pile surfaces of the primary loop (e. g. heat exchanger) which becomes a safety concern for maintenance personnel. Further, the flow of the medium replenishes the concentration and pH at the corroding site which aggravates the corrosion (flow-assisted corrosion).

An important aspect in the degradation of the Zircaloy clads, the sole bar­rier between the hot fuel and coolant, is the oxidation and corrosion problem. The various oxidation processes in Zircaloys have led to major degradation phenomena that are described in detail in subsequent chapters. Stable, adher­ent oxide films form which act as a protective coating and offer resistance to environmental cracking as in the case of stainless steels. However, in the case of Zircaloys long exposures lead to the film flaking — that results in wall thinning — which in some cases may result in through-wall failures. More and specific details can be found in the chapters on Zr-alloys in Part II.

The addition of transition metals (Fe, Cr, Ni) to zirconium was aimed at reducing the severe oxide growth stresses and localized cracking of the grain boundaries that exposes more grains to the corrosive environment. Formation of second phase precipitates (SPPs) containing the transition elements help by aiding uniform oxidation of the grains and preventing localized cracking and spalling as the oxide grows.3739 It has been noted in Zircaloy-4 that the fraction of SPPs decrease and correspondingly the oxide thickness increases with fluence (Fig. 1.27).40

A few factors under irradiation conditions are recognized as affecting the corrosion of clad material: (1) The increase in thickness of the oxide film (water side) decreases the thermal conductivity of the film and increases the temperature of the zirconium matrix which results in a higher corrosion rate;41 (2) the changes that occur in the microchemistry of the Zircaloy-4 matrix due to irradiation are seen to accelerate weight loss (when com­pared with unirradiated material);3 942 (3) accumulation of lithium in the oxide layer reduces the protective nature of the oxide at the metal oxide interface and enhances the corrosion rate;43 and (4) high concentrations of zirconium-hydride are found in locations where the oxidation of Zircaloy is high.3144 The vast research work conducted over the years has led to some understanding of these problems: contrary to the expectation that irradiation-induced defects can cause breakdown of the protective film on Zircaloy clads,45 the examinations of irradiated clads did not show any evi­dence of oxide damage;4647 though the exact mechanism of lithium induced corrosion of zirconium is not clear, the general understanding of the lithium effect is that Li gets incorporated in solid solution in the ZrO2 and alters the vacancy concentration and distribution in the oxide film; since zirconium alloy corrosion proceeds by oxygen diffusion through the film, the increased number of vacancies should increase the diffusion and, hence, the corrosion rate; further, it has always been noted in irradiated Zircaloys that hydride density is high in locations where the oxide thickness is high; this clearly indi­cates that the cathodic hydriding and anodically favoured oxidation occur

independent of each other. Change in fabrication route, which can result in the second phase with a different composition, can reduce the susceptibility to nodular corrosion but can lead to increased uniform corrosion.

The most commonly encountered corrosion types in the nuclear reactor are uniform corrosion, nodular corrosion and shadow corrosion. Uniform corrosion, as the name suggests, is associated with uniform oxide thicken­ing and is commonly seen in PWRs and BWRs. Unlike the PWR environ­ment where dissolved hydrogen is present in the coolant and the oxide films remained uniform over a very large thickness, the intermetallics present in Zircaloy, under the BWR environment, promoted nodular corrosion. The mechanism of oxide layer growth on Zircaloy under an irradiation envi­ronment is complex. Uniform corrosion starts with low burnup and the thickness of the grey oxide layer increases with burnup and operating tem­perature. Unlike in BWRs where the outlet temperature and pressure are limited, PWRs can operate at higher outlet temperatures but with the risk of increased corrosion, and this effect is explicitly seen from the increased oxide thickness with the elevation of the fuel rod. Figure 1.2848 depicts the profile of the oxide thickness layer with the elevation in a typical PWR fuel rod, indicating the increased oxide layer thickness with increase in tempera­ture. The increased turbulence in the coolant close to the spacer grids (which increases the cooling efficiency) and their parasitic absorption of neutrons result in the suppression of clad oxidation at these locations whereas the fuel rod temperature along the length, and hence the oxide thickness, is fairly uniform under a BWR environment.49

A more crucial parameter than metal coolant temperature is the metal oxide interface temperature which is difficult to measure but may be calcu­lated with large uncertainty. A further complexity arises as the thermal con­ductivity reduces with burnup (due to penetration by the coolant into the porosity, cracking and spalling of the oxide films and crud deposition). The measurement of the thermal conductivity of the loose or non-adherent crud layers, which modifies the metal oxide interface temperature, is extremely difficult as the properties of subsequent layers deposited may not be the same. It is known that the thermal conductivity of the crud is higher than the zirconia layer or water or steam50 which in a way increases the heat transfer characteristics.

Uniform corrosion occurs in both PWRs and BWRs. The oxide that forms is uniform in thickness, consists of several different layers and depends on many factors such as initial SPP size, extent of cold work and irradiation, alloy and water chemistries, temperature, local thermohydraulics, etc. The microstructure of the Zircaloys used in BWRs is continuously evolving, leading to dissolution of the SPP and formation of small and thin patches of white oxide on the otherwise black uniform oxide layer, which thicken at an accelerated rate. The sensitivity of nodular corrosion can be related to the second phase particles present in the alloy, though the number of nodules may not bear a one-to-one relation with the number of particles. Nodular corrosion is encountered in BWRs and starts appearing after a few to 100 days from the start of operation and usually saturates at higher expo­sure times. Nodules, in general, do not form in Zircaloys with small SPP sizes (<0.1 pm) but initiate early in materials with large SPPs and grow at a decreasing rate with fluence. Figure 1.2951 shows the appearance of nodular corrosion on the fuel clad of a BWR fuel pin. The shape of nodular corro­sion can be lenticular or spherical and growth in Zircaloy-2 decreases at high burnups.52 The nodular corrosion problem can be eliminated (or delayed) by judiciously controlling the second phase particle sizes through appro­priate в quench treatment although this may enhance uniform corrosion.53 Nodules, whose thickness greatly exceeds the uniformly growing film, are prone to spalling and promote hydrogen pick-up. They can also be a cause of introducing zirconia particles to the coolant. Though PWR and WWER structures are not prone to the nodular corrosion attack, nodular corrosion can be a problem if steam forms at the oxide-coolant interface.54

There is another type of localized corrosion in Zircaloys: the enhanced in-reactor corrosion when Zircaloy is placed close to a noble metal (under BWR conditions it is stainless steel or a nickel alloy), and where Zircaloy ‘mimics’ the noble metal corrosion. This is termed as ‘shadow corrosion’.55 The oxide thickness is unusually large and often appears to be particularly dense and uncracked. This localized corrosion is a special case of crevice corrosion and is predominantly seen in BWR components, although there is no direct electrical contact between Zircaloy and the material producing the shadow effect. The oxidation of H2O2 to HO2+H+ at the noble metal surface is balanced by the regeneration of H2O2 on the ZrO2 surface and the coupling between the two metals (Zircaloy and nickel) is maintained by the ionic transport under a concentration gradient. The driving force is the potential difference between the two metals and radiolysis of water is required to sustain this reaction.55 Shadow corrosion is invariably noted in BWRs and not in PWRs where the coolant is high in hydrogen concentra­tion, which in turn reduces or eliminates galvanic potentials between dis­similar alloy components.

Environmentally assisted cracking is another manifestation of corrosion — related problems and is very often encountered in the steam and feed water piping as well as in condensate systems, RPV feed water nozzles and the secondary circuit of LWRs. This process is accelerated by stress (i. e. SCC) and neutron flux (i. e. IASCC). A typical fracture surface of IASCC is shown in Fig. 1.30.56 Attempts are being made to reduce IASCC. Figure 1.31 shows the effect of hydrogen injection into the BWR environment on IASCC of 304 SS. The mechanism of crack growth mitigation by hydrogen injec­tion could be explained by analyzing the corrosion potential of the system. The presence of molecules like H2 O2 and O2 increases the free corrosion potential which falls into the cracking range and hence the crack velocity is enhanced following the slip dissolution model and Faraday’s law. Whereas, when hydrogen is introduced into the environment it helps the recombi­nation of species and thus reduces the corrosion potential far below the cracking range.57

Austenitic stainless steels (e. g. blade sheathing in BWR) at high tem­perature and in a neutron-rich environment (>0.7 dpa), further influenced by higher oxygen levels in the water (BWR environment), exhibit IASCC. Other steels and nickel-base alloys also undergo IASCC at lower stress levels. Another aspect is that IASCC occurs in almost all materials and is known to occur in components at low stress levels. It is an expensive process to detect and repair the affected component. SCC is a major issue of PWR components like steam generator tubes, RPV penetrations, pressurizer noz­zles, etc. While SCC can be controlled by modifying the water chemistry and the composition of the alloy, that is by replacing components with those resistant to SCC (e. g. Alloy 690, 52, 152), IASCC is more complex. Though

both IASCC and IGSCC require external stress, temperature and dissolved oxygen in the water environment, the former is accelerated by neutron radiation. It is recognized that the influence of water chemistry becomes weaker and disappears at high doses (>50 dpa) suggesting that mechanical processes dominate chemical processes in IASCC.58 The major distinctions between IASCC and other environmental cracking phenomena (e. g. SCC) are that in the former (1) the microstructure is modified by fast neutrons with time and (2) the chemistry of the environment is altered by the ionizing radiation. However, the overall stability of water increases with increasing temperature and the yields of molecular decomposition products (H2 , O2 and H2O2) correspondingly reduced.59 Austenitic stainless steel is the major material that has been the subject of IASCC investigation as compared to other grades. In the case of in-core structures, radiolysis increases the elec­trochemical potential in that region where the SCC susceptibility is high. Among the various radiolytic products, H2O2 is the most concentrated spe­cies present in irradiated, aerated water which gives rise to high corrosion potential for stainless steels. However, the critical potential to mitigate SCC of irradiated materials has not yet been established.6061

Slow strain-rate tests have been carried out on type 304 stainless steel with prior thermal sensitization of the grain boundaries (to produce grain boundaries with chromium depletion) that show that the electrochemical potential of stainless steel increased significantly on irradiation in oxygen­ated water but decreased slightly in the hydrogen treated water. Though the mechanism is not fully understood, it is now realized that neither Cr depletion near grain boundaries22 nor RIS (of S or P) at the boundar­ies63 alone plays the detrimental role in IASCC. Further, the low stacking fault energy (SFE) of the matrix leads to localized deformation through dislocation channelling and irradiation has been found to accelerate the IASCC process.64 Studies done under simulated BWR environments to examine the susceptibility of four steels with varying SFE under irradia­tion showed that the one with the highest SFE exhibited good resistance to cracking whilst that with the lowest SFE was seen to be susceptible to cracking at all of the doses studied25 (Fig. 1.32). In 316 SS, the initiation and propagation of IASCC in a water environment depends on the dis­solved hydrogen and the stress required decreases with (a) increase in dissolved hydrogen and (b) decrease in the rate of straining.66 Corrosion problems are equally important in storage and disposal of nuclear wastes where long term safety and reversibility act as guidelines in designing the basic layout of a geological repository. Unlike conventional engineering structures, the ageing and degrading clad tubes should not only serve trou­ble free all through their service life but also maintain their integrity in repository conditions.67